JPH073472B2 - Treatment of used solvent - Google Patents
Treatment of used solventInfo
- Publication number
- JPH073472B2 JPH073472B2 JP63222100A JP22210088A JPH073472B2 JP H073472 B2 JPH073472 B2 JP H073472B2 JP 63222100 A JP63222100 A JP 63222100A JP 22210088 A JP22210088 A JP 22210088A JP H073472 B2 JPH073472 B2 JP H073472B2
- Authority
- JP
- Japan
- Prior art keywords
- solvent
- freeze
- solution
- treatment
- vacuum
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Expired - Fee Related
Links
Classifications
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
- G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
- G21F9/04—Treating liquids
- G21F9/06—Processing
- G21F9/08—Processing by evaporation; by distillation
Description
【発明の詳細な説明】 〔産業上の利用分野〕 本発明は使用済核燃料の再処理プロセス、及びスクラッ
プ核燃料の湿式回収プロセス等に利用可能な使用済溶媒
の処理法に関するものである。DETAILED DESCRIPTION OF THE INVENTION [Industrial field of use] The present invention relates to a spent solvent treatment method that can be used in a spent nuclear fuel reprocessing process, a scrap nuclear fuel wet recovery process, and the like.
一般に、使用済核燃料の再処理プロセス、及びスクラッ
プ核燃料の湿式回収プロセスにおいて、抽出工程で使用
する有機溶媒は放射線や酸の影響により劣化するため、
その劣化生成物を水酸化ナトリウム溶液、炭酸ナトリウ
ム溶液で取り除いた後、再使用している。Generally, in the reprocessing process of spent nuclear fuel and the wet recovery process of scrap nuclear fuel, the organic solvent used in the extraction step deteriorates due to the effects of radiation and acid.
The deteriorated product is reused after being removed with sodium hydroxide solution and sodium carbonate solution.
しかし、このような従来の方法では、 劣化度が進行した有機溶媒は再生不可能となり、処理
困難な放射性廃液となる。However, according to such a conventional method, the organic solvent whose degree of deterioration has progressed cannot be regenerated and becomes a radioactive waste liquid which is difficult to process.
ナトリウムを含む溶液は硝酸系の放射性廃液と混合さ
れた後減容し、ガラス固化処理、アスファルト固化処理
を行うが、その際多くのナトリウムを含むため減容の制
限を受け、固化処理の繁雑さの原因にもなる。The solution containing sodium is mixed with the radioactive waste liquid of nitric acid system, and then the volume is reduced, and then vitrification treatment and asphalt solidification treatment are performed.However, since a large amount of sodium is contained in the solution, the volume reduction is limited and the solidification treatment is complicated. It also causes
等の欠点があり、そのため可能な限りナトリウムを使用
しないプロセスと溶媒再生プロセスの開発が望まれてい
る。Therefore, it is desired to develop a process that does not use sodium as much as possible and a solvent regeneration process.
また、放射性廃液の処理工程では放射性物質の濃縮のた
めに蒸発缶を使用するが、除染効率が低く、材料腐食が
大きいという欠点があり、それ故除染効率が高く、材料
腐食のない処理プロセスの開発が望まれている。In the process of treating radioactive waste liquid, an evaporator is used to concentrate radioactive substances, but it has the disadvantages of low decontamination efficiency and large material corrosion, and therefore high decontamination efficiency and no material corrosion treatment. Process development is desired.
本発明は上記問題点を解決するためのもので、凍結真空
乾燥法を使用することにより、低温での操作により材料
腐食がなくなり、火災爆発等の危険がなく、安全性が高
く、可能な限りナトリウム含有物質を使用しないように
してアスファルト固化設備、ガラス固化設備の省略化、
簡略化を図ると共に、回収液の再利用と放射性廃液の減
容を図り、さらに除染効率の高い真空蒸溜法を溶媒再生
に用いることにより、溶媒の再利用と放射線廃溶媒の減
溶を図り、ソルトフリープロセスの可能な使用済溶媒の
処理法を提供することを目的とする。The present invention is to solve the above problems, by using the freeze-vacuum drying method, the material corrosion is eliminated by operation at low temperature, there is no danger of fire explosion, etc., high safety, as much as possible. Asphalt solidification equipment and glass solidification equipment can be omitted without using sodium-containing substances,
In addition to the simplification, the recovered liquid is reused and the volume of radioactive waste liquid is reduced.By using the vacuum distillation method with high decontamination efficiency for solvent regeneration, the solvent can be reused and the radioactive waste solvent can be dissolved. , A salt-free process capable of treating used solvent is provided.
本発明は、使用済核燃料の再処理プロセス、スクラップ
核燃料の湿式回収プロセスにおいて、溶媒洗浄工程の使
用済溶媒に凍結真空乾燥法を用いてｎ−ドデカンを蒸発
させて分離し、残渣として残ったリン酸トリ−ｎ−ブチ
ル、リン酸ジブチルを真空蒸留法により分離することを
特徴とする。The present invention, in a spent nuclear fuel reprocessing process and a scrap nuclear fuel wet recovery process, evaporates and separates n-dodecane by using a freeze-vacuum drying method as a spent solvent in a solvent cleaning step, and leaves phosphorus as a residue. It is characterized in that tri-n-butyl acidate and dibutyl phosphate are separated by a vacuum distillation method.
本発明は、使用済核燃料の再処理プロセスおよびスクラ
ップ核燃料の湿式回収プロセスにおいて、ナトリウムを
含む廃液を減少させ、アスファルト固化処理、ガラス固
化処理等のプロセスを合理化するために、劣化溶媒の洗
浄をなくし、代わりに真空凍結乾燥法と真空蒸溜法を組
合わせて劣化溶媒から劣化生成物を取り除き、また放射
性廃液を除染効率の高い真空凍結乾燥法で処理すること
により、放射性物質のほとんどを残渣として回収すると
共に、回収溶液は再利用し、廃液の減少と廃液処理の簡
略化を図り、さらにプルトニウム溶液、ウラン溶液を真
空凍結乾燥法で硝酸塩として回収し、これを熱分解して
酸化物とすることにより酸化物粉末製品を得ることがで
きる。The present invention eliminates the cleaning of a deteriorated solvent in order to reduce waste liquid containing sodium and to streamline processes such as asphalt solidification treatment and vitrification treatment in a spent nuclear fuel reprocessing process and a scrap nuclear fuel wet recovery process. Instead, by combining the vacuum freeze-drying method and the vacuum distillation method to remove the degraded products from the degraded solvent, and by treating the radioactive waste liquid by the vacuum freeze-drying method with high decontamination efficiency, most of the radioactive substances are converted into residues. At the same time as recovering, the recovered solution is reused to reduce the waste liquid and simplify the waste liquid treatment.Furthermore, the plutonium solution and uranium solution are recovered as nitrates by the vacuum freeze-drying method, and thermally decomposed to form oxides. Thereby, an oxide powder product can be obtained.
以下、実施例を図面を参照して説明する。 Hereinafter, embodiments will be described with reference to the drawings.
図は本発明の使用済燃料の処理法の一実施例を示してい
る。図中、は溶解槽、は溶媒抽出工程、は硝酸プ
ルトニウム溶液、硝酸ウラニル溶液、は凍結真空乾燥
装置、は硝酸塩、は凝縮液、は脱硝工程、は焼
結還元工程、は製品、は使用済溶媒、は凍結真空
乾燥装置、はTBP,DBP等、はｎ−ドデカン、は真
空蒸溜装置、はDBP等、はTBP、は調製工程、は
焼却炉、は廃液、は凍結真空乾燥装置、は残渣、
は水，硝酸、は保管又は固体廃棄物処理系、は調
製工程、は利用工程、は放出工程である。The figure shows one embodiment of the method for treating spent fuel of the present invention. In the figure, is a dissolution tank, is a solvent extraction step, is a plutonium nitrate solution, uranyl nitrate solution, is a freeze vacuum dryer, is a nitrate, is a condensate, is a denitration step, is a sintering reduction step, is a product, is a used Solvents are freeze-vacuum dryers, TBP, DBP, etc. are n-dodecane, vacuum distillation devices, DBP, etc. are TBP, are preparation steps, incinerators, waste liquids, freeze-vacuum dryers, are residues ,
Is water, nitric acid, is a storage or solid waste treatment system, is a preparation process, is a utilization process, is a discharge process.
図において、燃料製造工場等で発生した不純物を含む核
燃料スクラップは硝酸溶液と共に溶解槽に供給され、
ここで加熱、溶解される。そしてプルトニウム溶液、ウ
ラン溶液は調製されてから溶媒抽出工程に送られ、TB
P、ｎ−ドデカン等からなる溶媒と硝酸溶液等を用い、
硝酸プルトニウム，硝酸ウラニル溶液と、使用済溶媒
と、廃液とに分けられる。In the figure, nuclear fuel scrap containing impurities generated at a fuel manufacturing plant etc. is supplied to a dissolution tank together with a nitric acid solution,
It is heated and melted here. The plutonium solution and uranium solution are prepared and then sent to the solvent extraction step, where TB
Using a solvent composed of P, n-dodecane, etc. and a nitric acid solution,
It is divided into plutonium nitrate, uranyl nitrate solution, used solvent, and waste liquid.
硝酸プルトニウム溶液，硝酸ラウニル溶液は、凍結真
空乾燥工程で硝酸塩と凝縮液に分離され、凝縮液
は凍結真空乾燥装置へ送られる。一方、硝酸塩は
脱硝工程へ送られ、例えばマイクロ波加熱して酸化物
にしてから焙焼還元炉等を用いた焙焼還元工程で必要
に応じて粉末調製して製品となる。The plutonium nitrate solution and the lauryl nitrate solution are separated into nitrate and condensate in the freeze vacuum drying process, and the condensate is sent to the freeze vacuum dryer. On the other hand, the nitrate is sent to a denitration step, for example, is microwave-heated to be an oxide, and then powdered if necessary in a roasting reduction step using a roasting reduction furnace or the like to obtain a product.
使用済溶媒は凍結真空乾燥装置で、TBP,DBP等と
ｎ−ドデカンとに分けられる。TBP,DBP等は真空蒸
留装置でDBP等とTBPとに分離され、DBP等は焼
却炉へ送られ、一方TBPとｎ−ドデカンとは調製
工程で混合され、さらに必要に応じてTBP,n−ドデカ
ン等を加えて調製後、溶媒抽出工程へ送られる。The used solvent is divided into TBP, DBP, etc. and n-dodecane by a freeze vacuum drying device. TBP, DBP, etc. are separated into DBP, etc. and TBP by a vacuum distillation apparatus, DBP, etc. are sent to an incinerator, while TBP and n-dodecane are mixed in the preparation process, and if necessary, TBP, n- After preparation by adding dodecane and the like, the mixture is sent to the solvent extraction step.
廃液は凍結真空乾燥装置へ送られ、プルトニウム、
ウラン、アメリシウム等の不純物よりなる残渣と水，
硝酸とに分けられる。残渣（硝酸塩）は回収のた
め、工程で保管または固体廃棄物処理系へ送られる。
水、硝酸は調製工程において、必要に応じて水，硝
酸を加え、もしくは濃縮したり希釈したりして調製し、
工程で利用され、例えば溶解槽、溶媒抽出工程、
その他、図示していない、例えばオフガス洗浄工程等に
送られる。もしその余裕が生じた場合には工程で放出
される。The waste liquid is sent to a freeze vacuum dryer, where plutonium,
Residue consisting of impurities such as uranium and americium, and water,
Divided into nitric acid. The residue (nitrate) is stored in the process or sent to a solid waste treatment system for recovery.
Water and nitric acid are prepared by adding or concentrating or diluting water or nitric acid as necessary in the preparation process,
Used in processes such as dissolution tanks, solvent extraction processes,
In addition, it is sent to, for example, an off-gas cleaning step (not shown). If there is a margin, it will be released in the process.
なお、上記実施例においては、凍結真空乾燥装置を，
，の３台用いるようにしているが、貯槽を設けて運
転するようにすれば、勿論凍結真空乾燥装置は１台でも
良い。In the above embodiment, the freeze vacuum drying device is
, 3 are used, but if a storage tank is provided to operate, of course, only one freeze-vacuum drying device may be used.
以上のように本発明によれば、溶媒洗浄工程に凍結真空
乾燥法を用い、TBP、DBP等とｎ−ドデカンを分離でき、
さらに溶媒洗浄工程に真空蒸溜法を用い、TBPとDBP等を
分離することができ、ナトリウム塩の使用をなくすこと
ができる。そのため、放射性廃液発生量が減少し、処理
を簡略化でき、放射性廃液の中和・濾過が不要となると
共に、スラッジ発生量を少なくすることができる。ま
た、放射性廃液を除染効率の高い凍結真空乾燥法で処理
することにより、放射性物質のほとんどを残渣として回
収すると共に、回収溶液は再利用し、廃液の減少と廃液
処理の簡略化を図り、さらにプルトニウム溶液、ラウン
溶液を連結真空乾燥法で硝酸塩として回収し、これを熱
分解して酸化物とすることにより酸化物粉末製品を得る
ことができる。As described above, according to the present invention, it is possible to separate TBP, DBP and the like from n-dodecane by using the freeze vacuum drying method in the solvent washing step,
Furthermore, by using the vacuum distillation method in the solvent washing step, TBP and DBP can be separated, and the use of sodium salt can be eliminated. Therefore, the amount of radioactive waste liquid generated is reduced, the treatment can be simplified, neutralization and filtration of the radioactive waste liquid are not required, and the amount of sludge generated can be reduced. In addition, by treating the radioactive liquid waste by a freeze-vacuum drying method with high decontamination efficiency, most of the radioactive substances are recovered as a residue, and the recovered solution is reused to reduce the amount of liquid waste and simplify the liquid waste treatment. Further, an oxide powder product can be obtained by recovering the nitrate solution from the plutonium solution and the laurne solution as a linked vacuum drying method and thermally decomposing the nitrate.
図は本発明の使用済燃料の処理法の一実施例を示す図で
ある。 ……溶解槽、……溶媒抽出工程、……硝酸プルト
ニウム溶液、硝酸ウラニル溶液、……凍結真空乾燥装
置、……硝酸塩、……凝縮液、……脱硝工程、
……焙焼還元工程、……製品、……使用済溶媒、
……凍結真空乾燥装置、……TBP,DBP等、……ｎ−
ドデカン、……真空蒸溜装置、……DBP等、……T
BP、……調製工程、……焼却炉、……廃液、…
…凍結真空乾燥装置、……残渣、……水，硝酸、
……保管又は固体廃棄物処理系、……調製工程、…
…利用工程、……放出工程。FIG. 1 is a diagram showing an embodiment of a method for treating spent fuel according to the present invention. …… Dissolution tank …… Solvent extraction process …… Plutonium nitrate solution, Uranyl nitrate solution …… Freezing vacuum drying device …… Nitrate …… Condensate …… Denitration process
…… Roasting reduction process …… Products …… Spent solvent,
...... Freeze-vacuum dryer, ...... TBP, DBP, etc. ・ ・ ・ n-
Dodecane, ... Vacuum distiller, ... DBP, etc .... T
BP, ... Preparation process, ... Incinerator, ... Waste liquid, ...
… Freeze vacuum drying equipment …… residues …… water, nitric acid,
…… Storage or solid waste treatment system …… Preparation process ……
… Utilization process… Release process.
───────────────────────────────────────────────────── フロントページの続き (56)参考文献 特開 昭62−27697（ＪＰ，Ａ) 特開 昭62−49296（ＪＰ，Ａ) 特開 昭54−23900（ＪＰ，Ａ) 特開 昭56−115991（ＪＰ，Ａ) ─────────────────────────────────────────────────── ─── Continuation of the front page (56) Reference JP 62-27697 (JP, A) JP 62-49296 (JP, A) JP 54-23900 (JP, A) JP 56- 115991 (JP, A)
Claims (1)
プ核燃料の湿式回収プロセスにおいて、溶媒洗浄工程の
使用済溶媒に凍結真空乾燥法を用いてｎ−ドデカンを蒸
発させて分離し、残渣として残ったリン酸トリ−ｎ−ブ
チル、リン酸ジブチルを真空蒸留法により分離すること
を特徴とする使用済溶媒の処理法。1. In a spent nuclear fuel reprocessing process and a scrap nuclear fuel wet recovery process, n-dodecane is separated by evaporating and separating n-dodecane by using a freeze vacuum drying method as a spent solvent in a solvent washing step. A method for treating a used solvent, characterized in that tri-n-butyl phosphate and dibutyl phosphate are separated by a vacuum distillation method.
Priority Applications (4)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP63222100A JPH073472B2 (en) | 1988-09-05 | 1988-09-05 | Treatment of used solvent |
| US07/400,220 US4981616A (en) | 1988-09-05 | 1989-08-29 | Spent fuel treatment method |
| DE68916135T DE68916135T2 (en) | 1988-09-05 | 1989-09-04 | Process for treating spent fuel. |
| EP89308938A EP0358431B1 (en) | 1988-09-05 | 1989-09-04 | Spent fuel treatment method |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP63222100A JPH073472B2 (en) | 1988-09-05 | 1988-09-05 | Treatment of used solvent |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPH0269697A JPH0269697A (en) | 1990-03-08 |
| JPH073472B2 true JPH073472B2 (en) | 1995-01-18 |
Family
ID=16777138
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP63222100A Expired - Fee Related JPH073472B2 (en) | 1988-09-05 | 1988-09-05 | Treatment of used solvent |
Country Status (4)
| Country | Link |
|---|---|
| US (1) | US4981616A (en) |
| EP (1) | EP0358431B1 (en) |
| JP (1) | JPH073472B2 (en) |
| DE (1) | DE68916135T2 (en) |
Families Citing this family (8)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPH0833485B2 (en) * | 1990-04-11 | 1996-03-29 | 動力炉・核燃料開発事業団 | Separation and purification method of spent solvent generated from nuclear fuel cycle |
| JPH0495899A (en) * | 1990-08-14 | 1992-03-27 | Power Reactor & Nuclear Fuel Dev Corp | Extraction and separation of spent solution generated from nuclear fuel cycle |
| JP2529457B2 (en) * | 1990-10-01 | 1996-08-28 | 動力炉・核燃料開発事業団 | Low temperature concentration method of plutonium nitrate solution |
| JP2551683B2 (en) * | 1990-10-01 | 1996-11-06 | 動力炉・核燃料開発事業団 | Method for separating uranium and plutonium from uranium-plutonium mixed solution |
| US5707592A (en) * | 1991-07-18 | 1998-01-13 | Someus; Edward | Method and apparatus for treatment of waste materials including nuclear contaminated materials |
| JP5067700B2 (en) * | 2009-02-23 | 2012-11-07 | 独立行政法人日本原子力研究開発機構 | Method for producing metal oxide particles |
| CN109830324B (en) * | 2019-01-17 | 2022-11-25 | 中国辐射防护研究院 | Feed liquid suitable for treating radioactive organic waste liquid by pyrolysis incineration method and preparation method |
| CN111863301B (en) * | 2020-06-10 | 2022-08-19 | 中国原子能科学研究院 | Method for eluting plutonium reserved in PUREX process waste organic phase |
Family Cites Families (15)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| GB994156A (en) * | 1962-04-27 | 1965-06-02 | Leybold Anlagen Holding Ag | Process for treating radioactive substances |
| DE1220048B (en) * | 1960-10-21 | 1966-06-30 | Leybold Hochvakuum Anlagen | Process for transferring radioactive substances into a permanent form that can be stored and transported |
| DE1199192B (en) * | 1962-01-13 | 1965-08-19 | Leybold Hochvakuum Anlagen | Process for drying goods under a porous top layer |
| US3361649A (en) * | 1965-04-05 | 1968-01-02 | American Mach & Foundry | Method and apparatus for distillation of waste liquids and separate recovery of solvent and solute |
| US3725293A (en) * | 1972-01-11 | 1973-04-03 | Atomic Energy Commission | Conversion of fuel-metal nitrate solutions to oxides |
| US4043936A (en) * | 1976-02-24 | 1977-08-23 | The United States Of America As Represented By United States Energy Research And Development Administration | Biological denitrification of high concentration nitrate waste |
| DE2728469C2 (en) * | 1977-06-24 | 1986-01-16 | Josef 5000 Köln Stecker | Method and device for the treatment of liquids containing radioactive waste |
| JPS5423900A (en) * | 1977-07-25 | 1979-02-22 | Mitsubishi Metal Corp | Recovering regeneration method of radioactive retreating waste organic solvent |
| JPS54121442A (en) * | 1978-03-13 | 1979-09-20 | Power Reactor & Nuclear Fuel Dev Corp | Microwave heating device for radioactive material |
| US4225455A (en) * | 1979-06-20 | 1980-09-30 | The United States Of America As Represented By The United States Department Of Energy | Process for decomposing nitrates in aqueous solution |
| JPS56115991A (en) * | 1980-02-19 | 1981-09-11 | Tokyo Shibaura Electric Co | Microwave heating deniration device |
| JPS5924738B2 (en) * | 1980-12-16 | 1984-06-12 | 株式会社東芝 | Nuclear fuel conversion device |
| JPS5930652B2 (en) * | 1981-04-16 | 1984-07-28 | 株式会社東芝 | Microwave heating denitrification equipment |
| JPS6227697A (en) * | 1985-07-29 | 1987-02-05 | 動力炉・核燃料開発事業団 | Method and device for processing waste liquor containing radioactive substance |
| JPS6249296A (en) * | 1985-08-28 | 1987-03-03 | 株式会社東芝 | Evaporating concentrator |
-
1988
- 1988-09-05 JP JP63222100A patent/JPH073472B2/en not_active Expired - Fee Related
-
1989
- 1989-08-29 US US07/400,220 patent/US4981616A/en not_active Expired - Lifetime
- 1989-09-04 DE DE68916135T patent/DE68916135T2/en not_active Expired - Fee Related
- 1989-09-04 EP EP89308938A patent/EP0358431B1/en not_active Expired - Lifetime
Also Published As
| Publication number | Publication date |
|---|---|
| DE68916135T2 (en) | 1994-09-22 |
| JPH0269697A (en) | 1990-03-08 |
| DE68916135D1 (en) | 1994-07-21 |
| EP0358431A1 (en) | 1990-03-14 |
| EP0358431B1 (en) | 1994-06-15 |
| US4981616A (en) | 1991-01-01 |
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